SOME OUTSTANDING PROBLEMS IN NEUTRON TRANSPORT COMPUTATION

This article provides selects of outstanding problems in computational neutron transport, with some suggested approaches thereto, as follows: i) ray effect in discrete ordinates method, ii) diffusion synthetic acceleration in strongly heterogeneous problems, iii) method of characteristics extension to three-dimensional geometry, iv) fission source and k eff convergence in Monte Carlo, v) depletion in Monte Carlo, vi) nuclear data evaluation, and vii) uncertainty estimation, including covariance data.

[1]  Richard B. Lehoucq,et al.  Krylov subspace iterations for the calculation of K-Eigenvalues with sn transport codes , 2002 .

[2]  D. Smith.  A unified Monte Carlo approach to fast neutron cross section data evaluation. , 2008 .

[3]  Ser Gi Hong,et al.  Acceleration and Parallelization of the Method of Characteristics for Lattice and Whole-Core Heterogeneous Calculations , 2000 .

[4]  G. I. Bell,et al.  Nuclear Reactor Theory , 1952 .

[5]  Ser Gi Hong,et al.  Extensions of Streaming Rays Method for Streaming Dominant Neutron Transport Problems , 1996 .

[6]  S. Woolf,et al.  Particle Transport Calculations with the Method of Streaming Rays , 1981 .

[7]  Gerardo Aliberti,et al.  Use of Covariance Data to Select Experiments Relevant to Target Systems for GNEP , 2008 .

[8]  A Rigorous Treatment of Transverse Buckling Effects in Two-Dimensional Neutron Transport Computations , 1980 .

[9]  Masayuki Nakagawa,et al.  Error estimations and their biases in Monte Carlo eigenvalue calculations , 1997 .

[10]  N. Cho,et al.  Coarse-Mesh Angular Dependent Rebalance Acceleration of the Method of Characteristics in x-y Geometry , 2008 .

[11]  Nam-Zin Cho,et al.  Refinement of the 2-D/1-D Fusion Method for 3-D Whole-Core Transport Calculation , 2002 .

[12]  Ser Gi Hong,et al.  CRX: A code for rectangular and hexagonal lattices based on the method of characteristics , 1998 .

[13]  E. M. Gelbard.  Biases in Monte Carlo eigenvalue calculations , 1992 .

[14]  Jeffrey C. Davis,et al.  Comparison of monte carlo and deterministic depletion codes for LWR fuel cycle analysis , 2005 .

[15]  Forrest B. Brown.  WIELANDT ACCELERATION FOR MCNP5 MONTE CARLO EIGENVALUE CALCULATIONS , 2007 .

[16]  Dorothea Wiarda,et al.  Low-fidelity Covariance Project , 2008 .

[17]  Y. Lee,et al.  NEUTRON CROSS SECTION DATA LIBRARY FOR PD-105, AG-109, XE-131 AND CS-133 , 2005 .

[18]  M. S. Moore,et al.  MULTILEVEL FORMULA FOR THE FISSION PROCESS , 1958 .

[19]  V. Weisskopf,et al.  Theoretical Nuclear Physics , 1953 .

[20]  Nam Zin Cho,et al.  A Fusion Technique of 2-D/1-D Methods for Three-Dimensional Whole-Core Transport Calculations , 2002 .

[21]  Jonghwa Chang,et al.  Measurement of neutron total cross-section of Dy at Pohang Neutron Facility , 2003 .

[22]  Jong Kyung Kim,et al.  A generalized perturbation theory program for CANDU core analysis , 2001 .

[23]  Carlos Alberto Brayner de Oliveira Lira,et al.  Recent advances in perturbative methods applied to nuclear engineering problems , 1998 .

[24]  W. H. Reed,et al.  Ray-Effect Mitigation Methods for Two-Dimensional Neutron Transport Theory , 1977 .

[25]  Nam Zin Cho,et al.  FUNDAMENTALS AND RECENT DEVELOPMENTS OF REACTOR PHYSICS METHODS , 2005 .

[26]  Toshikazu Takeda,et al.  Estimation of Error Propagation in Monte-Carlo Burnup Calculations , 1999 .

[27]  Elmer E Lewis,et al.  Ray-effect mitigation in discrete ordinate-like angular finite element approximations in neutron transport , 1975 .

[28]  N. Z. Cho,et al.  Fusion of method of characteristics and nodal method for 3-D whole-core transport calculation , 2002 .

[29]  J. Chang,et al.  NEUTRON CROSS SECTION EVALUATIONS OF FISSION PRODUCTS BELOW THE FAST ENERGY REGION , 2000 .

[30]  Yang-Hyun Koo,et al.  PROGRESS IN NUCLEAR FUEL TECHNOLOGY IN KOREA , 2009 .

[31]  A. Gandini,et al.  A generalized perturbation method for bi-linear functionals of the real and adjoint neutron fluxes , 1967 .

[32]  E. Gadioli,et al.  Pre-Equilibrium Nuclear Reactions , 1992 .

[33]  Edward W. Larsen,et al.  Fast iterative methods for discrete-ordinates particle transport calculations , 2002 .

[34]  Yoshinori Miyoshi,et al.  Fission Source Convergence of Monte Carlo Criticality Calculations in Weakly Coupled Fissile Arrays , 2000 .

[35]  Jim E. Morel,et al.  Effect of Material Heterogeneity on the Performance of DSA for Even-Parity S n Methods , 1999 .

[36]  N. García-Herranz,et al.  Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations , 2008 .

[38]  Y. Y. Azmy.  Impossibility of unconditional stability and robustness of diffusive acceleration schemes , 1998 .

[39]  Donald L. Smith.  Summary of the Workshop on Neutron Cross Section Covariances , 2008 .

[40]  E. Lewis,et al.  Computational Methods of Neutron Transport , 1993 .

[41]  G. E. Whitesides,et al.  DIFFICULTY IN COMPUTING THE K-EFFECTIVE OF THE WORLD. , 1971 .

[42]  Jaejoo Ha,et al.  OPPORTUNITIES AND CHALLENGES OF NEUTRON SCIENCE AND TECHNOLOGY IN KOREA , 2009 .

[43]  M. Salvatores,et al.  A Global Approach to the Physics Validation of Simulation Codes for Future Nuclear Systems , 2008 .

[44]  Yongdeok Lee,et al.  NEUTRON INDUCED CROSS SECTION DATA FOR IR-191 AND IR-193 , 2006 .

[45]  M. W. Herman,et al.  EMPIRE: Nuclear Reaction Model Code System for Data Evaluation , 2007 .